1. Field of the Invention
The present invention relates generally to fuel assemblies for a nuclear reactor and, more particularly, is concerned with a boiling water reactor (BWR) fuel assembly incorporating features which improve its structural integrity, hydraulic effects and critical heat flux (CHF) characteristics.
2. Description of the Prior Art
Typically, large amounts of energy are released through nuclear fission in a nuclear reactor with the energy being dissipated as heat in the elongated fuel elements or rods of the reactor. The heat is commonly removed by passing a coolant in heat exchange relation to the fuel rods so that the heat can be extracted from the coolant to perform useful work.
In nuclear reactors generally, a plurality of the fuel rods are grouped together to form a fuel assembly. A number of such fuel assemblies are typically arranged in a matrix to form a nuclear reactor core capable of a self-sustained, nuclear fission reaction. The core is submersed in a flowing liquid, such as light water, that serves as the coolant for removing heat from the fuel rods and as a neutron moderator. Specifically, in a BWR the fuel assemblies are typically grouped in clusters of four with one control rod associated with each four assemblies. The control rod is insertable within the fuel assemblies for controlling the reactivity of the core. Each such cluster of four fuel assemblies surrounding a control rod is commonly referred to as a fuel cell of the reactor core.
A typical BWR fuel assembly in the cluster is ordinarily formed by a N by N array of the elongated fuel rods. The bundle of fuel rods are supported in laterally spaced-apart relation and encircled by an outer tubular channel having a generally rectangular cross-section. Examples of such fuel assemblies are illustrated and described in U.S. Pats. Nos. 3,689,358 to Smith et al and 3,802,995 to Fritz el al and Canadian Pat. No. 1,150,423 to Anderson et al, as well as in the patent applications cross-referenced above.
In a fuel assembly of this type the fuel rods in the central region of the bundle thereof may be undermoderated and overenriched. In order to remedy this condition by increasing the flow of moderator water through this region of the assembly, several arrangements have been proposed. In the Fritz et al patent, one or more elongated empty rods are substituted for fuel rods in the central region of the assembly. In the above cross-referenced Olson et al patent application, water tubes are arranged in a cross-like pattern among the fuel rods in the assembly. In the Anderson et al patent, an elongated centrally-disposed stiffening device with vertical water passageways is used in the assembly. In the above cross-referenced Barry et al, Doshi and Lui patent applications, an elongated centrally-disposed water cross is used in the assembly.
As disclosed in the aforementioned latter three cross-referenced applications, the water cross has a plurality of four radial panels composed by a plurality of four, elongated, generally L-shaped metal angles or sheet members that extend generally along most of the length of the fuel assembly and are spaced apart by a series of interconnecting elements. The interconnecting elements are formed in the sheet members of each panel so as to extend therebetween in aligned, contacting pairs which are connected together such as by welding to ensure that the spacing between the sheet members of each panel is accurately maintained. Thus, the central water cross being in the form of a cruciform water flow channel divides the fuel assembly into four, separate elongated compartments, with the bundle of fuel rods being divided into mini-bundles disposed in the respective compartments, and provides a centrally-disposed cross-shaped path for the flow of subcooled neutron moderator water within the channel along the lengths of, but separated from, adjacent fuel rods in the mini-bundles thereof.
In the central water cross of the above cross-referenced Barry et al and Doshi applications, it was considered advantageous to provide some means to permit coolant flow between the separate minibundles of fuel rods of the fuel assembly to provide hydraulic pressure equalization therebetween. In Barry et al, coolant flow was permitted between fuel rod mini-bundles via openings defined by the pairs of interconnecting elements, whereas in Doshi such elements were changed to the form of closed interconnecting dimples and coolant flow was instead provided via openings in outer longitudinal ribs extending between and interconnecting the outer channel and the radial ends of the water cross panels. In both instances, however, it was felt necessary to prevent the mixing of near boiling coolant flowing within the fuel rod mini-bundles with subcooled moderator flowing through the central water cross.
Notwithstanding the improvements fostered by the cruciform water cross flow channel of the above cross-referenced Barry et al and Doshi applications with respect to hydraulic pressure equalization between the fuel rods mini-bundles, other problems have recently been recognized which, if left uncorrected, detrimentally affect the fuel rod bundle of the assembly and the structural integrity of the water cross. These problems generally related to an inadequate critical heat flux (CHF) margin in some axial regions along the fuel rod bundle and the occurrence of high static pressure loads which can cause water cross dimple weld failure.
With respect to CHF problems of the BWR fuel assembly, it is generally known that CHF occurs when the liquid film cooling a heated surface dries up, or gets washed away leaving a dry path. Under BWR operating conditions, this typically occurs in the annular flow regime. The location of greatest likelihood for a dry heated surface to result lies towards the exit of the heated region of the assembly.
With respect to structural and hydraulic problems of the BWR fuel assembly, conventionally the water cross inlet is designed with flow restrictions such that approximately 9 to 10 percent of the total bundle coolant flow enters the water cross. However, in view that there is little or no flow restriction at the outlet of the water cross, the largest means stresses (static pressure difference between mini-bundle and water cross) occur at its inlet region. Hence, the possibility of fatigue loading induced failure is greatest in the inlet region of the water cross. Indeed, failures of the dimple welds holding the water cross panels together have occurred at the inlet during testing.
By way of example, one water cross design known heretofore has inlet holes of approximately 0.14 inch diameter size while there is no restriction at its outlet. These inlet holes have small cross-sectional areas, and crud buildup during operation can decrease this area. Small area reductions can lead to a substantial increase in orificing (hydraulic loss) which, in turn, leads to even larger pressure losses at the inlet. Hence, lesser flow would tend to enter the water cross leading to the possibility of moderator boiling as well as increased static pressure loading. That is, the fluid static pressure difference between the mini-bundles and the water cross would increase, thus worsening the situation. Heretofore, fluid pressure in the mini-bundles has been greater than that in the water cross at all axial positions. Thus, if at some point fatigue loading or some other mechanism(s) causes dimple weld failure, fluid would enter the water cross from the mini-bundle region. The amount of flow being diverted into the water cross would depend on several factors (e.g. width of opening, pressure differences, etc.); however, this diversion causes depletion of fluid in the mini-bundles, leading to a degradation of fuel rod cooling and a reduction in CHF margin. The consequences of such a situation could mean undesirable operational penalties.
Consequently, the need exists for further improvement of the BWR fuel assembly so as to prolong its useful life by signficantly improving its CHF characteristics and enhancing its structural integrity.